1. Field of the Invention
The present invention relates to a "wet" oxidation process for reducing the volume of hazardous, radioactive, and mixed wastes, and for converting said wastes into a form suitable for storage, particularly long-term storage in a repository. More particularly, the present invention relates to a process for treating waste containing both organic carbon compounds and radioactive or hazardous material to reduce the volume of the material by oxidizing the organic carbon compounds with a combination of nitric acid and phosphoric acid, and then converting the reduced volume waste material into an immobilized final form, such as a glass or ceramic, which can then be stored in a suitable repository.
2. Description of Background and Related Art
The disposal of radioactive, hazardous, and mixed (radioactive and hazardous) waste has over the years become a growing environmental, political, and economic problem. Due to the limited number and capacity of suitable repositories and the political difficulties involved in establishing new repositories, the supply of disposal capacity has decreased. At the same time, increasing amounts of waste material must be disposed of due to nuclear disarmament, increasing awareness of existing waste in short term storage, and the production of new waste material in areas such as nuclear power plant operation and medical research.
A particular area of concern is the disposal of low level radioactive and mixed wastes, such as job control waste (i.e., waste generated by everyday operations in nuclear facilities such as protective gloves, clothing, etc. worn by workers who handle or are possibly exposed to radioactive material), nuclear power plant operations (such as contaminated solutions and ion exchange resins used to remove corrosion from reactor secondary cooling systems), and operations involving treatment and purification of water used to cool stored nuclear material, such as fuel rods (e.g., ion exchange resins). At the present time, over 80% of this type of waste is sent for storage to a single site, which is at or near capacity. As a result of this lack of available storage capacity and the measures taken by political entities to limit the amount of waste storage, costs to store low level waste have increased significantly. In order to reduce these costs, attempts have been made to reduce the amount of waste that must be sent to repositories. One method proposed has been to eliminate some or all of the components of low level waste that are not hazardous or radioactive, and/or convert hazardous components to nonhazardous form.
An additional concern with the storage of any radioactive or hazardous waste is the stability of the final storage form. Such waste must be safely stored for time periods that are often geological in scale, requiring that the material be stored in a form that is stable over time and also over exposure to a variety of conditions. The tendency of storage containers to break down or corrode over time and the resulting risk that the stored material will escape into the biosphere has led to the use of storage forms wherein the waste materials are immobilized in a solid form that is relatively stable toward the expected environments to which the stored material may be exposed. Immobilizing the waste material in a glass (vitrification) or ceramic that is stable over time to the conditions expected to be encountered in a repository are two examples of this approach.
Prior attempts to reduce the volume of hazardous or radioactive waste have involved several different approaches, some of which also involve immobilizing the radioactive material in a solid form.
U.S. Pat. No. 3,957,676 (Cooley et al.) describes treating combustible solid radioactive waste materials with concentrated sulfuric acid at a temperature within the range of 230.degree. C.-300.degree. C., and simultaneously and/or thereafter contacting the reacted mixture with concentrated nitric acid or nitrogen dioxide, in order to reduce the volume of combustible material and convert it into gaseous products.
U.S. Pat. No. 4,039,468 (Humblet et al.) describes an approach of attempting to separate radioactive species using solvent extraction. An organic phosphate-containing solvent is contacted with the waste and then treated by contacting the stream with phosphoric acid, obtaining a light organic phase containing essentially no radioactive material, and heavy aqueous and organic phases which contain essentially all of the radioactive material. The light organic phase can then be combusted, and the concentrated radioactive material can be solidified by reaction on aluminum oxide and incorporation into a glass or resin matrix.
U.S. Pat. No. 4,460,500 (Hultgren) describes reducing the volume of radioactive waste, such as ion exchange resins, by treatment with an aqueous complex forming acid, such as phosphoric acid, citric acid, tartaric acid, oxalic acid, or mixtures thereof to remove the radioactive species from the exchange resins and form a complex therewith. The radioactive species are then adsorbed onto an inorganic sorbent. The resulting material is then dried and calcined in the presence of air or oxygen, resulting in combustion of the organic material. The calcinated material is then collected into a refractory storage container, which is then heated to a temperature at which the material sinters or is fused to a stable product.
U.S. Pat. No. 4,732,705 (Laske et al.) describes treating radioactive ion exchange resin particles with an additive containing anions or cations that reduce the swelling behavior of the resin particles and produces a permanent shrinkage of the resin particles. The additive may be a polysulfide or organic acid ester. The treated resin particles are then immobilized in a solid matrix, such as a cement.
U.S. Pat. No. 4,770,783 (Gustavsson et al.) describes decomposing organic ion exchange resins containing radioactive materials by oxidation in a mixture of sulfuric acid and nitric acid in the presence of hydrogen peroxide or oxygen as an oxidant. Radioactive metals in the resulting liquid are precipitated with hydroxide and separated from the liquid, which contains other non-radioactive materials. The liquid is then released to the environment. The precipitated metal compounds are immobilized in cement.
U.S. Pat. No. 4,904,416 (Sudo et al.) describes centrifuging wet radioactive ion exchange particles to remove water therefrom, then coating the particles with a small quantity of cement powder, and then adding water and cement, in order to increase the loading of resin in the cement. U.S. Pat. No. 5,424,042 (Mason et al.) also describes removing water from radioactive ion exchange resins prior to vitrification.
U.S. Pat. No. 5,457,266 (Bege et al.) suggests dewatering radioactive ion exchange resins by mixing with a calcium compound and heating to a temperature over 120.degree. C. at a pressure of 120 hPa to 200 hPa.
These attempts have not been completely successful because (1) the use of sulfuric acid and other acids to oxidize organic materials included in waste streams does not allow for efficient conversion of the resulting treated waste stream into a stable, immobilized final form, (2) processes involving one or more transfers of radioactive species between solvent or sorbent phases is complicated and inefficient, (3) dewatering and cementation processes do not result in sufficient volume reduction, and (4) processes using high temperatures are not viewed favorably by the nuclear industry for oxidation of materials containing organic compounds.
Prior attempts to immobilize low level radioactive or mixed waste, such as ion exchange resins, have also been made.
U.S. Pat. No. 4,483,789 (Kunze et al.) describes a method for encasing the radioactive ion exchange resin in blast furnace cement. The mixture of resin, cement, and water is disclosed to have a slow initial hardening and high sulfate resistance, and is allowed to harden at room temperature.
U.S. Pat. No. 4,530,723 (Smeltzer et al.) describes a method for forming a solid monolith by mixing radioactive ion exchange resin and an aqueous mixture of boric acid or a nitrate or sulfate salt, a fouling agent, a basic accelerator, and cement, and allowing the cement to harden.
U.S. Pat. No. 4,632,778 (Lehto et al.) describes a process for disposing of radioactive material by adsorbing the radioactive material on an inorganic ion exchanger, mixing the inorganic ion exchanger loaded with radioactive species with a ceramifying substance and baking this mixture to form a ceramic.
U.S. Pat. No. 4,834,915 (Magnin et al.) describes immobilizing radioactive ion exchange resins by saturating them with a base, preferably sodium hydroxide and immobilizing them in a hydraulic binder. U.S. Pat. No. 4,892,685 (Magnin et al.) describes immobilizing radioactive ion exchange resins by first treating them with an aqueous solution containing NO.sub.3.sup.- and Na.sup.+ ions to ensure that all of the sites in the resin are saturated, and then adding a hydraulic binder, such as cement. U.S. Pat. No. 5,143,653 (Magnin et al.) describes treating borate containing radioactive ion exchange resins with calcium nitrate prior to incorporation into a hydraulic binder. These three patents are directed to attempting to resolve the problem of ion exchange of radioactive material between the immobilized resin material and the hydraulic binder.
U.S. Pat. No. 5,288,435 (Sachse et al.) describes a process for the incineration and vitrification of radioactive waste materials, which may contain sulfur compounds, by contact of the waste materials with molten glass in a glass melter having an extended heated plenum to allow for sufficient combustion residence times. If sulfur-containing wastes are being processed, the off gases produced can be scrubbed of sulfur, which can then be converted into gypsum.
U.S. Pat. No. 5,435,942 (Hsu) describes treating alkaline radioactive wastes with nitric acid to reduce pH and with formic acid to remove mercury compounds, in order to adjust the glass forming feedstock composition to achieve more efficient glass melter operation.
The use of lead-iron phosphate glasses for the immobilization of radioactive waste is described in U.S. Pat. Nos. 4,847,008 and 4,847,219 (Boatner et al.). The use of glasses to immobilize radioactive waste is also described in U.S. Pat. No. 3,161,601 (Barton), U.S. Pat. No. 3,365,578 (Grover), U.S. Pat. No. 4,351,749 (Ropp), and U.S. Pat. No. 5,461,185 (Forsberg et al.).
These methods of immobilizing radioactive materials are disadvantageous because the volume reduction of waste is inadequate, which results in increased costs for disposing of the organic, non-radioactive materials. In addition, removal of radioactive material is incomplete. Finally, any significant volume reduction that occurs is due to incineration, which creates the risk that radioactive species will be entrained in ash in the off gas.
It is an object of the present invention to avoid the disadvantages of the prior procedures by providing a simple, efficient process for the wet oxidation of organic carbon-containing radioactive, hazardous, or mixed waste products. It is also an object of the present invention to provide a process that results in significant volume reduction of these waste materials, thereby significantly decreasing the costs associated with their long term disposal. It is also an object of the present invention to provide a process whereby the residual concentrated waste product produced by the wet oxidation process is conveniently and easily incorporated into a final form material without special intermediate treatment steps. Finally, it is also an object of the present invention to provide a process for immobilizing radioactive, hazardous, or mixed waste products in a final form that is stable to expected repository conditions over long periods of time.